KR101668727B1 - Method for treatment of spent radioactive ion exchange resins, and the apparatus thereof - Google Patents

Abstract

The present invention relates to a method and a device for processing spent ion exchange resin including radionuclide and, more specifically, to a method and a device for processing spent ion exchange resin including radionuclide through a phased heat treating process. The present invention is provided to process the spent ion exchange resin including the radionuclide by using a method for volatilizing the radionuclide without gasifying carbon which is a main component, and separating/collecting the radionuclide through the phased heat treating process, thereby preventing problems of depositing a pollutant in an exhaust gas process caused by volatilizing volatile radionuclide including radioactive Cs or Sr, discharging the same to the air, generating exhaust gas including SO_2 and SO_3 having high concentration (more than thousands of ppm) discharged in a case of separating an ion exchanging device, and generating global warming gas CO_2. The present invention obtains a maximal volume reducing effect by condensing the volatile radionuclide and fixing/treating same.

Description

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method and apparatus for treating spent ion exchange resins containing radioactive nuclides,

The present invention relates to a method and an apparatus for treating waste ion exchange resin containing a radionuclide, and more particularly, to a method for treating a waste ion exchange resin containing a radionuclide through a stepwise heat treatment process and an apparatus for performing the same .

The basic concept of liquid radioactive waste disposal in light water reactor type nuclear power plant currently operating in Korea is evaporated and concentrated using a waste liquid evaporator, and then subjected to solidification treatment using paraffin as a solidifying agent. However, various types of foreign substances present in the radioactive waste liquid have a problem that the performance of the evaporator is reduced due to formation of bubbles and scale in the evaporator, the operation cost is increased due to frequent maintenance, and the continuous operation is hindered. In order to solve these problems, researches on a waste solution treatment method using an ion exchange resin, a precise or ultrafiltration membrane, and a reverse osmosis membrane have been actively carried out in overseas cases.

Among them, the ion exchange resin used for the treatment of the radioactive waste liquid is a polymer material which is made by bonding an ion exchanger to a polymer gas having a fine three-dimensional structure and exchanging and purifying an ionic substance dissolved in a polar and nonpolar solution , And the movable ion contained in the ion exchange resin replaces the other ions in the solution to purify the waste liquid. These ion exchange resins have been widely used for separation, purification and decontamination in various water treatment processes. In particular, in the nuclear industry, ion exchange resins are used to purify clean water and cooling water of nuclear power plants. On the other hand, such an ion exchange resin occurs as a waste ion exchange resin when the use time is long. Since the waste ion exchange resin thus produced is made of a polymer polymer having excellent mechanical strength and chemical resistance and the functional group on the surface thereof is replaced with a radionuclide and remains in an ion exchange form, It is one of the most difficult-to-handle flammable radioactive waste generated in nuclear power plants because it is very difficult to isolate and extract attached radionuclides.

On the other hand, waste ion exchange resin used in the purification process at the nuclear power plant has a relatively low level of radioactive waste, and the condensate purification plant of CPP (Condensate Polishing Plant) wastewater and steam generator extraction system There are BD (blowdown) effluent generated during the purification, and the effluent discharged from the liquid waste purification process in the Liquid Radwaste System (LRS) is a slightly higher effluent level. Generally, about 5,000 ~ 7,000 liters of waste water per year is generated per year, and the waste water thus generated is placed in a drum or a carbon steel drum according to the radioactive level and temporarily stored in a power plant for permanent disposal. Normally, when waste water from a nuclear power plant detects even a small amount of radioactivity, the whole amount is stabilized and sent to a radioactive waste repository in a drum. These radioactive wastes can not be classified as general wastes unless the radioactivity is reduced to a radioactive level capable of self-disposal. Particularly, when there is leakage of the steam generator tubule, the waste water generated at the purification of the extraction water system at the secondary side of the steam generator is treated with radioactive carbon (C-14), radioactive cesium (Cs-137) and radioactive cobalt Co-60), and such contaminated wastes can not be classified as general wastes because they are not completely extinguished even after long-term storage.

Therefore, the development of technologies to separate and extract radioactive nuclides to such an extent that self-disposal is possible from such contaminated wastewater is very important to reduce radioactive waste disposal and disposal costs.

Of the conventional treatment methods for treating such waste ion exchange resins, Korean Patent Registration No. 10-2008-0087360 discloses a method for increasing the treatment amount by increasing the amount of radioactive waste ion exchange resin to be treated, Exchange resin dewatering device. Specifically, a multipurpose container containing a radioactive waste ion-exchange resin and having pores formed in its bottom surface; An intermediate container for receiving the multi-purpose machine to protect the multi-purpose machine; And a shielding container for containing the intermediate container and shielding radiation emitted from the radioactive waste ion exchange resin.

However, in the case of the above-mentioned invention, the volume of the radioactive waste resin after use can be reduced, but the exhaust gas containing SO ₂ and SO ₃ and CO _{2, which} is a global warming gas, are generated in the process.

On the other hand, as another method for treating the conventional waste ion exchange resin, an inorganic component including residual nuclides, that is, ash (ash) is pyrolyzed or oxidized by gasification and treatment of a combustible organic component by a method such as incineration, And the like.

Korean Patent No. 10-2002-0031822 discloses an apparatus and a process for incinerating and melting a radioactive waste, for example, a waste ion exchange resin is treated using the method of incineration and vitrification as described above. And more particularly, to an apparatus and a process for treating exhaust gas generated in a process of incinerating and melting vitrifiable low-level radioactive waste by vitrification.

However, the above-mentioned incineration and vitrification treatment method has an advantage of providing the maximum calming effect by leaving the radionuclide with other inorganic substances, that is, a trace amount of ash components in the organic components. However, not only does it exhaust large amounts of CO ₂ , which is a global warming gas, but also harmful gases such as sulfur dioxide, unburned hydrocarbons, dioxins and nitrogen oxides (NOx) as well as high-temperature volatile radionuclides (Cs-137, Cs-134), and the like.

That is, the conventional waste incineration and vitrification process is such that the waste ion exchange resin generated in a nuclear-related facility is mostly composed of an organic component, that is, a combustible component, and can effectively reduce the volume through incineration or pyrolysis. The radioactive nuclei fixed in the waste ion exchange resin in the process may be discharged together with the exhaust gas.

Accordingly, it is urgently required to develop a method for treating spent ion exchange resin, which solves the problem of sulfur oxides and carbon dioxide generated during waste ion exchange resin treatment, and effectively captures radionuclides and provides a sweetening effect.

Accordingly, the present inventors have found that a method of treating a waste ion exchange resin containing a radionuclide, wherein a radioactive nuclide is volatilized and recovered without gasification of carbon, which is a main component of a waste ion exchange resin, through a stepwise heat treatment method (Cs) or volatile radionuclides (Sr) such as strontium (Sr), which is a problem that has arisen in conventional incineration and vitrification treatments, by treating waste ion exchange resins containing radionuclides The problem of the generation of exhaust gas containing SO ₂ and SO ₃ at a high concentration (several thousand ppm or more) and the generation of CO _{2, which} is a global warming gas, And at the same time, the radionuclide is volatilized and separated and recovered through the immobilization treatment, thereby obtaining the maximum absorbing effect And the present invention has been completed.

Korean Patent No. 10-2008-0087360 Korea Patent No. 10-2002-0031822

It is an object of the present invention to provide a method for treating a waste ion exchange resin comprising a radionuclide and an apparatus for performing the same.

Drying the spent ion exchange resin comprising the radionuclide (step 1);

Separating the ion exchanger containing radionuclides in the dried waste ion exchange resin (step 2);

Converting the compound comprising volatile radionuclides produced from the separated ion exchanger into a sulfur oxides containing a non-volatile radionuclide (step 3);

Converting the sulfuric acid containing the radionuclide into a chloride containing a radionuclide (step 4); And

Separating and recovering the radionuclide from the chloride containing the radionuclide by volatilization and condensation (step 5);

And a radioactive nuclide containing the radioactive nuclide.

Further, according to the present invention,

 A dryer (100) for drying a waste ion exchange resin comprising a radionuclide;

A screw conveyor reactor 400 for stepwise heat-treating the dried waste ion exchange resin discharged from the dryer;

And an inorganic chlorination reactor (600) for converting reactants discharged from the screw conveyor reactor into chlorides. The present invention also provides an apparatus for treating waste ion exchange resin comprising a radionuclide.

The present invention relates to a method for treating a waste ion exchange resin containing a radionuclide, and more particularly, to a method for separating and recovering a radionuclide by volatilizing it without gasification of carbon as a main component of a waste ion exchange resin through a stepwise heat treatment method to be. By treating the waste ion exchange resin containing radionuclides through the stepwise heat treatment method as described above, volatile radionuclides such as radioactive cesium (Cs) or strontium (Sr), which are problems in conventional incineration and vitrification treatments, are volatilized The problem of the emission of exhaust gas containing SO ₂ and SO ₃ at a high concentration (several thousand ppm or more) generated when the ion exchanger is separated, and the problem of generation of CO ₂ , which is a global warming gas, And at the same time, the radionuclide is volatilized and separated and recovered through the immobilization treatment, thereby achieving the maximum suppression effect.

BRIEF DESCRIPTION OF THE DRAWINGS FIG. 1 is a process diagram schematically showing a process for treating a waste ion exchange resin containing a radionuclide according to the present invention,
FIG. 2 is a schematic diagram showing separation of a sulfonic acid group (SO ₃ H ⁺ ) in the spent ion exchange resin and a functional group including Cs and Sr radionuclides,
3 is a graph showing a form of a compound which can be formed according to temperature and oxygen partial pressure in a Cs-OS system,
4 is a graph showing a temperature and a thermodynamic equilibrium concentration for achieving chloride conversion of Cs ₂ SO ₄ according to step 4 of Example 1 of the present invention,
5 is a graph showing the temperature and the thermodynamic equilibrium concentration for the conversion of SrSO ₄ to chloride according to Step 4 of Example 1 of the present invention,
6 is a graph showing the temperature and the thermodynamic equilibrium concentration for achieving chloride conversion of BaSO ₄ according to step 4 of Example 1 of the present invention.

According to the present invention,

Drying the spent ion exchange resin comprising the radionuclide (step 1);

Separating the ion exchanger containing radionuclides in the dried waste ion exchange resin (step 2);

Converting the compound comprising volatile radionuclides produced from the separated ion exchanger into a sulfur oxides containing a non-volatile radionuclide (step 3);

Converting the sulfuric acid containing the radionuclide into a chloride containing a radionuclide (step 4); And

Separating and recovering the radionuclide from the chloride containing the radionuclide by volatilization and condensation (step 5);

And a radioactive nuclide containing the radioactive nuclide.

Hereinafter, the method for treating a waste ion exchange resin containing a radionuclide according to the present invention will be described in detail.

In the method for treating a waste ion exchange resin containing a radionuclide according to the present invention, step 1 is a step of drying a waste ion exchange resin containing a radionuclide.

This is a step for gasifying and removing waste ion exchange resins H ₂ O and CO ₂ when heated so that dissolved solids and suspended solids contained in the waste ion exchange resin may be concentrated.

The waste ion exchange resin may be generated by using ion exchange resin for a long time for the purpose of separation and purification in various water treatment processes including the nuclear industry. That is, the waste ion exchange resin may be an ion exchange resin used in various fields such as production of pure water, high purity water, treatment of wastewater, recovery of valuable materials, medicines, food refining, Or a waste ion exchange resin containing a radionuclide, which is used to purify wastewater containing a radionuclide.

The waste ion exchange resin includes a functional group in which an ion exchange resin is chemically bonded to a polymer gas to ionize the ion exchange resin. The waste ion exchange resin includes a functional group formed by substituting the ion included in the tolerable group and the radionuclide in the waste solution It may be a waste ion exchange resin. For example, the ion exchange resin may include a functional group containing a sulfonic acid group (SO ₃ H +) in a styrene divinylbenzene copolymer gas, and a cation in the ion exchange resin and a radioactive nuclear species A waste ion exchange resin containing radioactive nuclides may be formed by substitution with each other.

For example, the ion exchange resin may contain a cation such as a hydrogen ion to remove a radionuclide containing Cs in the nuclear wastewater, wherein if there is one Cs ion bearing a charge of 2+ from the surrounding solution , Which can be substituted with two hydrogen ions charged to +1 and become a waste ion exchange resin.

At this time. The step of drying the waste ion exchange resin is a step of removing moisture contained in the waste ion exchange resin, and may be performed at a temperature range of 100 to 150 ° C. At this time, when the temperature for performing the drying of the waste ion exchange resin in step 1 is less than 100 ° C, vaporization of moisture is not performed, and when the temperature exceeds 150 ° C, separation of the ion exchanger occurs, There is a possibility that the gas containing SO ₂ as well as moisture may be discharged at the same time.

In the method of treating a waste ion exchange resin comprising a radionuclide according to the present invention, step 2 is a step of separating an ion exchanger containing a radionuclide in the dried waste ion exchange resin.

This is because the separation of the ion exchanger is for separating the radionuclide from the spent ion exchange resin, and the radionuclide separated in step 2 may remain in volatile oxide or hydroxide form.

Hereinafter, the process of separating the ion exchanger of step 2 according to the present invention is described in detail with reference to FIG.

As shown in FIG. 2, the waste ion exchange resin containing the radioactive nucleus includes a styrene divinylbenzene polymer and a functional group containing a sulfonic acid group (SO ₃ H ⁺ ) and a radionuclide cation (M ⁺ ). group) (may include an SO ₃ M ^+), in this case, the ion exchange group is a sulfonic acid group (SO ₃ H +) or a radionuclide cation (functional group (functional group) is SO ₃ M ⁺ days, including M ⁺⁾ .

The ion exchanger in the waste ion exchange resin is separated through the step 2, and after separation, a gas including a styrene divinylbenzene copolymer, an oxide including a radioactive nuclide, and a sulfurous acid gas (SO ₂ ) may be generated.

At this time, the separation of the ion exchanger of step 2 can be performed at a temperature range of 150 to 400 ° C.

If the temperature for carrying out the separation of the ion exchanger in step 2 is less than 150 ° C, there may arise a problem that the separation of the ion exchanger in the spent ion exchange resin is not appropriately performed. Further, if the temperature for performing the removal of the ion exchanger in step 2 exceeds 400 ℃, sulfur dioxide (SO ₂₎ The generation speed increases because the shorter the residence time in the reactor, the sulfur dioxide (SO ₂₎ and can cause also a problem which is the radionuclide, and SO ₂ gas contacts the unreacted do not smoothly performed occurrence of the radionuclide.

At this time, the sulfur dioxide gas (SO ₂ ) generated is sufficiently retained in the reactor, and reacted with the radionuclide component in the waste ion exchange resin to convert it into sulfuric acid containing a radionuclide. Further, in step 2, since decomposition of the organic component does not occur and only a small amount of sulfur dioxide (SO ₂ ) which is not reacted with the radionuclide is discharged, unlike the conventional incineration or vitrification process of waste ion exchange resin, SO ₂ and SO ₃ There is no possibility of generating a large amount of exhaust gas contained therein, so that a large-capacity exhaust gas processing apparatus may not be required.

Further, when a sulfur dioxide gas (SO ₂ ) exists in a slight amount on the gas phase, the radionuclide can be converted into sulfur oxides. Therefore, by leaving the sulfur dioxide (SO ₂ ) generated in the step 2 without releasing it to the outside, Radionuclides may also be converted to sulfur oxides containing non-volatile radionuclides.

As an example of this, FIG. 3 shows the form of a compound which can be formed according to temperature and oxygen partial pressure in the Cs-OS system. This is the case of such as shown for the compounds in the form of in the atmosphere within the oxide of CsO ₄ is present is sulfur dioxide (SO ₂₎ containing the radioactive nuclides, as shown in FIG. 3, radioactive cesium (Cs), the temperature or oxygen (O ₂ can be converted to non-volatile sulfuric acid form Cs ₂ SO ₄ when the partial pressure of sulfur dioxide (SO ₂ ) on the gas is at least 10 ^-15 atm.

In the method for treating a waste ion exchange resin comprising a radionuclide according to the present invention, step 3 is a step of converting a compound containing a volatile radionuclide generated from the separated ion exchanger into a sulfur oxide containing a non-volatile radionuclide .

At this time, the conversion to the sulfuric acid in the step 3 may be performed at a temperature range of 400 to 550 ° C.

If the temperature for the conversion of the step 3 to the sulfuric acid is less than 400 ° C, the conversion of the compound containing the radionuclide to the sulfuric acid is so slow that the radionuclides may not be completely converted to the sulfuric acid . If the temperature for carrying out the conversion to the sulfuric acid in step 3 exceeds 550 ° C, there may arise a problem that the compound containing the radionuclide can be gasified and discharged before being converted to the sulfur oxide.

For example, in the case of cesium, when it exceeds 550 ° C, Cs ₂ O, CsOH and Cs ₂ O ₂ H ₂ Gasification and so on.

The substituted radionuclides in the spent ion exchange resin may be at least one selected from the group consisting of Cs, Sr, Mn, Fe, Ba, Ni and Co, _{2 SO 4, SrSO 4, BaSO} 4, NiSO 4, FeSO 4, MnSO 4 , and CoSO _4, and the like. At this time, the sulfur oxides containing the radionuclides are nonvolatile, and the radionuclides are not gasified at a temperature of 700 ° C or lower.

In addition, the step (2) may further comprise the step of forming residual organic matter generated from the spent waste ion exchange resin into a carbide. At this time, the carbonization of the residual organic material can proceed more satisfactorily at 550 to 700 ° C, but is not limited thereto.

The residual organic component may include carbon, oxygen, hydrogen, and nitrogen components. The residual organic component including oxygen, hydrogen, and nitrogen may be vaporized in the temperature range of 550 to 700 ° C, It may be formed of a carbide.

Through the step 3 and the step of forming a carbide, the compound containing the volatile radionuclide can be converted into the sulfur oxide containing the non-volatile radionuclide, and the volatile radionuclide can be converted into the sulfur oxide containing the volatile radionuclide, Traces of hydrogen, nitrogen and oxygen components in the residual organic components involved can be gasified and the carbon component can be carbonized. At this time, since the radionuclide and the sulfur dioxide (SO ₂ ) are not contained in the gasified component, problems of deposition in the exhaust gas process by the conventional volatilization and discharge into the atmosphere together with the exhaust gas, It is possible to eliminate the problem of generating exhaust gas containing gas (SO ₂ ).

 In the method for treating a waste ion exchange resin comprising a radionuclide according to the present invention, step 4 is a step of converting the sulfuric acid containing the radionuclide into a chloride containing a radionuclide.

At this time, the conversion to the chloride containing radionuclides is preferably performed at 800 to 900 ° C.

If the temperature for carrying out the conversion of the step 4 to the chloride is lower than 800 ° C, there may arise a problem that the rate of chlorination is slowed down. Also, if the temperature for carrying out the conversion to the chloride of step 4 exceeds 900 DEG C, corrosion problems of the chlorination apparatus may occur.

The sulfur oxides containing the radionuclides are stable at high temperatures, but in the presence of chlorine gas, they can react with chlorine gas to convert them into chlorides.

For example, FIGS. 4 to 6 show that oxygen gas (O ₂ ) and chlorine gas (Cl ₂ ) exist in nitrogen (N ₂ ) or argon (Ar) ₂ SO ₄ , SrSO _4, and BaSO ₄ in an amount of 1 ppm by weight, respectively.

As shown in FIGS. 4 to 6, oxygen gas (O ₂ ) and chlorine gas (Cl ₂ ) are present in an amount of about 100 ppm each in an inert gas such as nitrogen (N ₂ ) or argon (Ar) Cs ₂ SO ₄ , SrSO ₄ and BaSO ₄ Are mixed at 1 ppm each, sulfur oxides containing radionuclides can be converted to chlorides at a temperature of about 800 ° C or higher.

On the other hand, at this time, in step 4, the carbon component is not chlorinated or gasified by the chlorine gas (Cl ₂ ). Accordingly, the carbon component in the high-temperature heat treatment at 1400 ° C. or more in the following step 5 can exist in the carbide form without being volatilized.

In the method for treating a waste ion exchange resin containing a radionuclide according to the present invention, step 5 is a step of separating and recovering the radionuclide from the chloride containing the radionuclide by volatilization and condensation.

In the step 5, the radionuclide can be condensed and volatilized after being volatilized, and the carbon component is separated into carbide. Thus, the radioactive nuclide can be maximally depressed when the radionuclide is discarded, and the generation of carbon dioxide can be minimized, The waste ion exchange resin can be treated in an environmentally friendly and efficient manner.

At this time, the volatilization of the radionuclide in step 5 may be performed at 1400 to 1500 ° C under a reduced pressure of about 1 Torr or less. If the temperature for volatilizing the radionuclide in step 5 is less than 1400 ° C, the volatilization of the radionuclide chloride may not be performed rapidly and the treatment time may become longer, If the temperature exceeds 1500 DEG C, there may occur a problem that unnecessary energy is consumed.

On the other hand, the volatile separated radionuclides can be subsequently immobilized by being condensed. In the case of carbides, since the radionuclides are not volatilized in step 5, the radionuclides remain after being volatilized, and the residual carbides do not contain radioactive materials , Recycled, or treated as general waste.

The present invention also relates to

A dryer (100) for drying a waste ion exchange resin comprising a radionuclide;

A screw conveyor reactor 400 for stepwise heat-treating the dried waste ion exchange resin discharged from the dryer;

And an inorganic chlorination reactor (600) for converting reactants discharged from the screw conveyor reactor into chlorides. The present invention also provides an apparatus for treating waste ion exchange resin comprising a radionuclide.

Hereinafter, a waste ion exchange resin treatment apparatus including a radionuclide of the present invention will be described in detail with reference to the drawings.

1, a waste ion exchange resin treatment apparatus including a radionuclide according to the present invention includes a dryer 100, a screw conveyor reactor 400, an inorganic chloride reactor 600, And a wet cleaning tower 700.

The dryer 100 according to the present invention is an apparatus for removing moisture in waste ion exchange resin and can be operated at a temperature of 100 to 150 ° C.

On the other hand, when waste ion exchange resin in the dryer drying, the water vapor (H ₂ O) separated by gasification with carbon dioxide (CO ₂₎ gas in addition can include tritium (H-3) and radiocarbon (C-14) And may further include a water condenser 200 and a carbon dioxide (CO ₂ ) absorption and recovery device 300 for cleaning and discharging the water condensed water. At this time, in order to easily discharge the water vapor and the carbon dioxide, which are separated from the dryer 100 by gasification, into the water condenser 200 and the carbon dioxide (CO ₂ ) absorption and recovery apparatus 300, N ₂ ) or argon (Ar) gas.

The screw conveyor reactor 400 according to the present invention is a device for stepwise heat treating waste ion exchange resin that has passed through the dryer 100 and can be divided into two or more regions having different temperatures. That is, the screw conveyor reactor 400 may be divided into an ion exchanger separation area 401 and a sulfur oxide conversion area 402, and may further include a carbide formation area 403.

The ion exchanger separation region 401 in the screw conveyor reactor 400 can be operated at a temperature of 150 to 400 ° C. and the heat treatment in the region can be performed using a sulfonic acid group (SO ₃ H ⁺ ) and a heat treatment zone for performing in order to remove the ion exchange group containing SO ₃ M ^+, this sulfonic acid group (SO ₃ H ⁺⁾ and SO ₃ M ⁺ is the ion-exchange separation, including a radionuclide oxide or hydroxide containing over And SO ₂ (SO ₂ ) can be generated.

The sulfur oxide conversion region 402 in the screw conveyor reactor 400 can be operated at a temperature of 400 to 550 ° C while being connected to the ion exchange group separation region 401. The heat treatment in the conversion region 402 into the sulfur oxides is a heat treatment region performed to form sulfur oxides including a radioactive nuclide to thereby convert the volatile radionuclides generated in the ion exchange cleavage region to nonvolatile , And the discharge of sulfur dioxide can be prevented.

In addition, the screw conveyor reactor 400 may further include a carbide forming region 403 followed by the sulfur oxide conversion region 402, and the carbide forming region 403 may have a temperature of 550 to 700 ° C Temperature can be operated. The heat treatment in the carbide formation region 403 is a heat treatment region in which oxygen, hydrogen, nitrogen, and the like contained in the residual organic components are discharged and carbonized to carbonize, and the following radioactive nuclides are volatilized and separated , It is possible to prevent gasification of the carbon component.

The screw conveyor reactor 400 may further include a heater including a heater outside the reactor. That is, most of the reactions occurring in the screw conveyor reactor 400 are endothermic reactions, and further include a heater including a heater outside the reactor, so that the temperature inside each reactor can be appropriately maintained.

The inorganic chlorination reactor 600 according to the present invention is an apparatus for forming a chloride including a radioactive nuclide in a reactant passing through the screw conveyor reactor 400. In the reactor, chlorine gas and oxygen gas are injected And can be operated at a temperature of 800 to 900 ° C.

Further, the chloride containing the radioactive nuclide formed through the inorganic chlorination reactor 600 can be separated and recovered by volatilizing the radioactive nuclide through heat treatment at 1400 to 1500 ° C. At this time, the wet scrubber 700 may be connected to the inorganic chlorination reactor, and the wet scrubber 700 may condense radionuclides volatilized through the inorganic chlorination reactor 600, The washing waste liquid in which the radionuclide species is concentrated through the centrifugal separator 700 may be dried and solidified to separate and recover the radioactive nuclides.

Hereinafter, the present invention will be described in detail with reference to the following examples.

However, the following examples are illustrative of the present invention and the scope of the invention is not limited by the examples.

≪ Example 1 >

Step 1: A waste ion exchange resin in which a Cs radionuclide in the ion exchange resin containing a sulfonic acid group (SO ₃ H ⁺ ) was substituted for styrene divinylbenzene copolymer gas was used. ) And dried at a temperature of 150 ° C for 2 hours and 30 minutes.

Step 2: The spent waste ion exchange resin dried in step 1 was placed in an ion-exchange separating zone 401 of a screw conveyor reactor 400 and heated to about 350 ° C. At about 150 ° C at the inlet, Lt; RTI ID = 0.0 > 400 C < / RTI > was maintained in this region. At this time, the generated sulfur dioxide gas can be gradually discharged through the conversion region to the next sulfur oxide.

Step 3: The reactants obtained from the step 2 were heat-treated at a temperature of about 550 ° C for 30 minutes in the conversion region 402 of sulfur oxides in the screw conveyor reactor 400.

Step 4: The reactants obtained from Step 3 were heat treated in a carbide forming region 403 in a screw conveyor reactor 400 at a temperature of about 700 캜 for 2 hours and then discharged.

Step 5: The reactants obtained in step 4 were placed in the inorganic chlorination reactor 600, and heat treatment was performed at a temperature of about 800 DEG C for about 90 minutes in an atmosphere of chlorine gas 100 ppm, oxygen gas 100 ppm, and nitrogen gas 1 atm.

Step 6: The reactants obtained from step 5 were heat treated in a vacuum atmosphere at about 1400 ° C, and the vaporized radionuclides were separated and sent to a wet scrubber to be condensed.

100: dryer
200: Water condensation recovery device
300: Carbon dioxide recovery device
400: screw conveyor reactor
401: Ion exchanger separation area
402: conversion area to sulfur oxides
403: Carbide forming region
500: SO ₂ adsorption / absorber
600: Inorganic Chlorination Reactor
700: Wet scrubber
800: high efficiency particulate air filter (HEPA) filter system

Claims (16)

delete delete delete delete delete delete delete delete delete delete delete A dryer for drying a waste ion exchange resin comprising a radionuclide;
A screw conveyor reactor for stepwise heat-treating the dried waste ion exchange resin discharged from the dryer;
And an inorganic chlorination reactor for converting a reactant discharged from the screw conveyor reactor into a chloride.
The apparatus for treating waste ion exchange resin according to claim 12, wherein the screw conveyor reactor is divided into an ion exchanger separation region and a sulfur oxide conversion region.
14. The apparatus of claim 13, wherein the screw conveyor reactor further comprises a carbide forming region followed by a conversion region to the sulfur oxides.
13. The apparatus for treating waste ion exchange resin according to claim 12, wherein the waste ion exchange resin further comprises a heater outside the screw conveyor reactor.
13. The waste ion exchange resin treating apparatus according to claim 12, further comprising at least one member selected from the group consisting of a water condenser, a carbon dioxide absorption recovery apparatus, and a wet cleaning tower. Ion exchange resin treatment apparatus.

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